1. Field
The field relates to quantitative safety analysis, in particular, to a method of determining safety factors in high risk non-stationary engineering systems and processes, such as a process of reloading nuclear fuel in a nuclear power plant (APP) and a control system and control method using thereof.
2. Description of Related Technology
Many current modelling systems are designed for performing probabilistic safety analysis solely and provide for the qualitative evaluation of failure probability of the engineering process. According to this approach, safety analysis was usually restricted to the brief characterisation of operating elements called “objects” and possible failures called “initial events”, and subsequent functional analysis of probability of one or the other process discontinuity or object damage for a small number of initiating events causing such discontinuity or damage.
Other methods use a deterministic approach which reviews the physical characteristics of a system, for example, temperature, pressure, etc., and evaluates the system solely on the basis of this quantitative information.
Some modelling systems combine a statistical and probabilistic approach to compare the present state of a component with its past history and to determine what could happen next. The current modelling methods do not emphasize an heuristic approach to consider the dynamic interaction between the components of a system or between the systems themselves when determining the present and future performance of a plant.
However, the growing complexity of engineering processes, in particular those connected with exploitation of thermoelectric power stations (TPS) and especially APP, due to a large number of logical and functional relations and time-dependence of safety parameters requires the development of new approaches to safety evaluation.
For example, U.S. patent application 20040086071 discloses an optimum evaluation system for safety analysis of a nuclear power plant, wherein data derived from results of a various kinds of experiments are used to improve codes so that the calculated results do not exceeds the experimental results at any condition, so that a sufficient safety margin is maintained at any condition. The system provides for quantification and standardization of the analysis method to three procedures. A first procedure relates to applying conditions and codes consists of a step for describing an accidental scenario, a step for selecting a subject power plant, a step for confirming main conditions and deciding the raking, a step for selecting an optimum code, a step for arranging documents related with the codes, and a step for deciding applicability of the codes. A second procedure evaluates the codes and deciding displacement of variables consists of: a step for evaluating codes and deciding evaluation matrix related to the displacement decision for the variables, a step for deciding nodding of a power plant, a step for deciding accuracy of the codes and the experiments, a step for analyzing and evaluating a scale effect decision, a step for deciding input variables of a nuclear reactor and their states related with the factors obtained by analyzing uncertainty and sensitivity, a calculating step of sensitivity of a power plant, a step for statistically evaluating uncertainty and a step for deciding a total uncertainty. A third procedure relates to analyzing sensitivity and evaluating uncertainty conducted by a step for evaluating bias which have not been considered in the first and the second procedures to decide a temperature of a final coating material.
This system allows estimating the safety of the existing objects only and cannot provide developing technical specification for the safety measures at the modernization and at the development of the new equipment for APP.
Attempts were made to reduce the problem of safety evaluation to selection of one possible decision of a plurality of decisions stored in a data base, which would be the most appropriate for the case. Thus, according to a method of a computer-aided safety analysis of a nuclear reactor (WO03/005376), functioning of APP is limited within the range of its safe exploitation, which is defined by the following steps:                a) providing the results of previously implemented safety analysis;        b) check up if the range of the safe exploitation of the APP defined earlier is applicable in the new operating conditions of the APP        
However, this method is applicable to safety analysis of only those APP's, which are already in operation and not to newly constructed or modified APP's.
U.S. Pat. No. 4,632,802 discloses a system for safety evaluation of APP, which provides continuous operation of APP in case of failure or unavailability of one or several APP elements. According to U.S. Pat. No. 4,632,802, the system provides for monitoring and evaluating the degree of risk associated with continued operation of a nuclear power plant while one or more plant components has failed or is otherwise unavailable. The apparatus has several functional sections, including means for storing a plant-specific data base of component-level core damage logic paths and component level failure probabilities, means for selecting plant condition scenarios by modifying the component failure probabilities to represent plant components actually or potentially unavailable, means for associating a figure of merit with the change in risk of core damage resulting from the unavailable components, and means for displaying the figure of merit relative to a base or reference value. As an interactive tool in the plant, the apparatus PSES displays the probability or risk of core damage almost instantly for any given state of plant readiness.
Similarly, the known system can be used for safety evaluation of existing APP and not for the newly developed or modified plants to optimize the APP equipment parameters, like a control system, to choose necessary and sufficient number of protection layers and of locks providing the object safety.
Another method for deterministic safety analysis based on the risk conception (EP1378916) includes ranking of initiating events depending on frequency of their occurrence, the threshold level of frequency of initiating events, acceptance criteria with adjustable level of conservatism, conservatism value using the methodology of the safety analysis, wherein the analysis of the events is performed using deterministic analysis in case the frequency of the event initiation exceeds the threshold level, or probabilistic analysis in case the frequency of the event initiation is below the threshold level.
The known method includes also the identification of the additional system of failures, which are not in a direct relation with initiating events, and definition of the common threshold frequency value for the combination of the initiating events frequency and additional failures frequency. Later the additional system of failures is appended to the safety analysis until the total frequency of event and additional failures does not exceed the threshold frequency level.
While the known method provides determining conditions when either deterministic or probabilistic method, it does not provide for the use of both methods when needed.
Further, EP 0411873 discloses a control system for a plant using a modelling system employing expert, deterministic and probabilistic modelling methods. This modelling system is implemented as a hierarchical structure of independent objects interacting with each other. Each object represents an element or a system. Objects are connected to each other through a data base available for all objects. The structure of the object module and the hierarchical structure are standardized and provide introducing new elements or systems by introduction of standard object modules including specific object model. The object model contains a deterministic model of the element degradation, probabilistic model of the element degradation and expert rules combining deterministic and probabilistic models with experts' knowledge aiming to determine the current state of the object and produce recommendations concerning future actions with respect to the object.
Further, according to international standards, a procedure for probabilistic safety analysis of APP in defined in Procedures for conducting probabilistic safety assessment of nuclear power plants (level 1), International Atomic Energy Agency, Vienna, 1992, STI/PUB/888. According to this standardised procedure, a probabilistic safety analysis of APP includes the following steps: input data acquisition and analysis, selection of input events, determining safety functions, determining functional system interconnection, determining successful functioning criteria, grouping input events, modelling a sequence of events and systems, and performing quantitative and qualitative safety analysis.
The above discussed approaches proved their effectiveness when applicable to stationary, in the context of safety conditions, systems only, wherein safety conditions either invariable or change relatively slowly, for example, as a result of ageing of system elements.
However, many complex engineering systems operate in safety conditions, which are both time and location-dependent and could vary within a single technological operation, as well as in course of a technological cycle, that makes the above discussed safety analysis approaches inapplicable.
The above said relates for example, to technological processes of transportation, which are typically characterized by significant changes in safety conditions both from one technological operation to the other, and within a single technological operation. There is a multitude of logical and functional inter-relations between operations that affect the overall safety of the engineering process.
As a result, known methods and systems become unreliable in case of safety analysis of complex engineering processes such as a process of reloading a nuclear fuel. The non-stationary character of such engineering processes makes impossible a reliable evaluation of safety conditions using the known methods and approaches.